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At least 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Evaluation and Application of Phyto-remediation at FUSRAP Projects - 16275

The Niagara Falls Storage Site (NFSS) and associated vicinity properties (VPs) occupy approximately 607 hectares (1,500 acres) of the original 3,035-hectare (7,500-acre) Lake Ontario Ordnance Works (LOOW). The NFSS was used to store radioactive materials, residues and waste from uranium ore processing between 1944 and 1952. A field-study is currently underway to evaluate the potential use of hybridized trees to stabilize radionuclides, mainly uranium, and chlorinated solvents derived from trichloroethylene degradation in groundwater at the NFSS. The purpose of this effort is to support the USACE-Buffalo District in the evaluation and implementation of phyto-remediation approaches to contain groundwater impacts on the site. This demonstration project is considered a site maintenance action and will have relevance to the Formerly Utilized Sites Remedial Action Program (FUSRAP) mission. The demonstration is designed to determine the ability of trees to withstand and uptake Uranium (U), and to a lesser extent Trichloroethylenes (TCEs), and hydraulically control groundwater where historical practices produced a series of disconnect plumes. Although the groundwater at the NFSS is naturally high in total dissolved solids that preclude untreated consumption, the groundwater contaminant levels are very high in some areas and may limit the options for future site re-use. Active groundwater remediation is impractical due to the clayey site soils and low risk of transport. The phyto-remediative approach is a passive action to lessen the groundwater impacts during the project's life cycle and property re-use planning. The target contaminant sources and plumes occur near the property boundaries, along legacy sanitary sewer lines, and near on-site surface-water courses. The demonstration includes soil characteristic determinations, the testing of optimal hybrid tree species, field implementation at target sites, and soil moisture and groundwater monitoring during the long-term evaluation phase. The final disposition of the trees will be included in future site decision and maintenance documents. (authors)

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NASA NTRS1974

Feasibility of space disposal of radioactive nuclear waste. 2: Technical summary

The feasibility of transporting radioactive waste produced in the process of generating electricity in nuclear powerplants into space for ultimate disposal was investigated at the request of the AEC as a NASA in-house effort. The investigation is part of a broad AEC study of methods for long-term storage or disposal of radioactive waste. The results of the study indicate that transporting specific radioactive wastes, particularly the actinides with very long half-lives, into space using the space shuttle/tug as the launch system, appears feasible from the engineering and safety viewpoints. The space transportation costs for ejecting the actinides out of the solar system would represent less than a 5-percent increase in the average consumer's electric bill.

Source record↗
DOE OSTI2015

Integrated Waste Treatment Unit (IWTU) Input Coal Analyses and Off-Gass Filter (OGF) Content Analyses

A full engineering scale Fluidized Bed Steam Reformer (FBSR) system is being used at the Idaho Nuclear Technology and Engineering Center (INTEC) to stabilize acidic Low Activity Waste (LAW) known as Sodium Bearing Waste (SBW). The INTEC facility, known as the Integrated Waste Treatment Unit (IWTU), underwent an Operational Readiness Review (ORR) and a Technology Readiness Assessment (TRA) in March 2014. The IWTU began non-radioactive simulant processing in late 2014 and by January, 2015 ; the IWTU had processed 62,000 gallons of simulant. The facility is currently in a planned outage for inspection of the equipment and will resume processing simulated waste feed before commencing to process 900,000 gallons of radioactive SBW. The SBW acidic waste will be made into a granular FBSR product (carbonate based) for disposal in the Waste Isolation Pilot Plant (WIPP).

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DOE OSTI2017

Nucleation and crystal growth behavior of nepheline in simulated high-level waste glasses

The Savannah River National Laboratory (SRNL) has been tasked with supporting glass formulation development and process control strategies in key technical areas, relevant to the Department of Energy’s Office of River Protection (DOE-ORP) and related to high-level waste (HLW) vitrification at the Waste Treatment and Immobilization Plant (WTP). Of specific interest is the development of predictive models for crystallization of nepheline (NaAlSiO4) in HLW glasses formulated at high alumina concentrations. This report summarizes recent progress by researchers at SRNL towards developing a predicative tool for quantifying nepheline crystallization in HLW glass canisters using laboratory experiments. In this work, differential scanning calorimetry (DSC) was used to obtain the temperature regions over which nucleation and growth of nepheline occur in three simulated HLW glasses - two glasses representative of WTP projections and one glass representative of the Defense Waste Processing Facility (DWPF) product. The DWPF glass, which has been studied previously, was chosen as a reference composition and for comparison purposes. Complementary quantitative X-ray diffraction (XRD) and optical microscopy confirmed the validity of the methodology to determine nucleation and growth behavior as a function of temperature. The nepheline crystallization growth region was determined to generally extend from ~ 500 to >850 °C, with the maximum growth rates occurring between 600 and 700 °C. For select WTP glass compositions (high Al2O3 and B2O3), the nucleation range extended from ~ 450 to 600 °C, with the maximum nucleation rates occurring at ~ 530 °C. For the DWPF glass composition, the nucleation range extended from ~ 450 to 750 °C with the maximum nucleation rate occurring at ~ 640 °C. The nepheline growth at the peak temperature, as determined by XRD, was between 35 - 75 wt.% /hour. A maximum nepheline growth rate of ~ 0.1 mm/hour at 700 °C was measured for the DWPF composition using optical microscopy. This research establishes a viable alternative to more traditional techniques for evaluating nepheline crystallization in large numbers of glasses, which are prohibitively time consuming or otherwise impractical. The ultimate objective is to combine the nucleation and growth information obtained from DSC, like that presented in this report, with computer simulations of glass cooling within the canister to accurately predict nepheline crystallization in HLW during processing through WTP.

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DOE OSTI2018

Crystalline Silicotitanate (CST) Ion Exchange Media Performance Evaluations in SRS Average Supernate Simulant and Tank 10H Waste Solution to Support TCCR

Savannah River Remediation (SRR) is deploying an “at-tank” ion exchange process to remove radioactive cesium from waste supernate which is referred to as the Tank Closure Cesium Removal (TCCR) system utilizing Crystalline Silicotitanate (CST) media. After being loaded with cesium, but before flushed and dried and sent to Interim Safe Storage (ISS), it is anticipated that the TCCR ion exchange columns will be taken off-line (stagnant) with the CST media still immersed in supernate. The currently designed columns include no active cooling system during processing (although no rmal process fluid flow through the column during operations provides significant cooling) or during off-line conditions and temperature increases are expected in the stagnant, loaded, and liquid-filled columns due to decay heat from the radionuclides absorbed to the CST. As a result, modification of the Tank 10H supernate composition has been considered through the addition of either NaOH or KOH reagents to decrease cesium loading.

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DOE OSTI2025

Colloidal Behavior of Plutonium Oxide in Concentrated Electrolyte Solutions

The Hanford Site in Washington State manages legacy high-level radioactive waste streams that display major chemistry and engineering challenges, including the high salt levels and pH values that correspond to conditions under which many classical concepts describing chemical reactivity cannot be applied. One particular challenge that needs to be tackled at Hanford is that Pu concentrations, [Pu], in the soluble phases of the tank wastes are higher than expected based on the solubility of crystalline PuO2, which is widely accepted to be caused by the formation of PuO2 colloid, consisting of nano- to submicron-sized particles (PuO2 NPs). Fundamental research underpinning the behavior of PuO2 NPs under conditions not only relevant to the Hanford tank waste but at high ionic strength in general is needed to reliably predict the chemical reactivity of PuO2 NPs and develop engineering solutions to safely and efficiently process high-level radioactive waste into forms suitable for long-term storage. In this work, we study the behavior of PuO2 NPs (particle size ~100 nm) under high ionic strength conditions by reacting it with highly concentrated (up to 5 M) salt solutions. We explore different electrolyte compositions to elucidate the impact of different anions (NO3-, Cl-, ClO4-, SO42-, C2O42-, CO32-) on the stability of PuO2 NP in the acidic and alkaline pH regime. PuO2 NP aggregation and precipitation as function electrolyte concentration is tracked by a combination of liquid scintillation counting, dynamic light scattering for determination of particle size distributions, and zeta potentials as a proxy for particle charge. At acidic pH, electrolytes containing non-coordinating anions, such as NaNO3, NaCl, and NaClO4 mostly stabilize PuO2 NPs over a large electrolyte concentration range, showing only subtle differences in their reactivity. Other electrolyte anions show a more pronounced effect on the PuO2 NP stability: SO42-, binds directly to the particles’ surface, reverses the particle charge, and precipitates the PuO2 NPs efficiently even at intermediate sulfate concentrations (>0.1 M). In contrast, C2O42- is found to lead to high [Pu] in solution, in the milli-molar range, even at mildly acidic pH (~4). Thermodynamic modeling of the dissolved Pu concentrations using PHREEQC is unable to predict the observed [Pu] in the acidic pH regime, supporting the influence of colloids in maintaining elevated [Pu]. It is noteworthy that the current thermodynamic databases do not include constants for colloidal Pu phases and cannot accurately predict many of the high ionic strength solutions relevant to this work. The mechanisms and models responsible for these observations will need further investigation in the future. At high pH values (~12), PuO2 NPs exhibits classical sol-gel chemistry, meaning that upon destabilization of the colloidal sol, for example by addition of concentrated NaOH, highly porous and viscid PuO2 coagulates are formed that consist of a three-dimensional network likely held together by physical interactions. The PuO2 NP coagulate shows no significant reversibility of the aggregation when contacted with concentrated brines; however, PuO2 NPs can be efficiently resuspended in solution by addition of diluted electrolytes, alkaline solutions containing high amounts of carbonate, or simple addition of water. Especially carbonate is shown to stabilize PuO2 NPs in solution at high pH, characterized by stable colloidal suspensions that are resistant against sedimentation during centrifugation. Thermodynamic modeling of the carbonate system was able to predict an increasing dissolved Pu concentration with increasing carbonate concentration. However, the model was profoundly sensitive to the fixed redox potential and does not include any thermodynamic constants for colloidal Pu species.

Neumann, Julia↗
DOE OSTI2016

Advances in Geologic Disposal System Modeling and Application to Crystalline Rock

The Used Fuel Disposition Campaign (UFDC) of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (OFCT) is conducting research and development (R&D) on geologic disposal of used nuclear fuel (UNF) and high-level nuclear waste (HLW). Two of the high priorities for UFDC disposal R&D are design concept development and disposal system modeling (DOE 2011). These priorities are directly addressed in the UFDC Generic Disposal Systems Analysis (GDSA) work package, which is charged with developing a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic media (e.g., salt, granite, clay, and deep borehole disposal). This report describes specific GDSA activities in fiscal year 2016 (FY 2016) toward the development of the enhanced disposal system modeling and analysis capability for geologic disposal of nuclear waste. The GDSA framework employs the PFLOTRAN thermal-hydrologic-chemical multi-physics code and the Dakota uncertainty sampling and propagation code. Each code is designed for massively-parallel processing in a high-performance computing (HPC) environment. Multi-physics representations in PFLOTRAN are used to simulate various coupled processes including heat flow, fluid flow, waste dissolution, radionuclide release, radionuclide decay and ingrowth, precipitation and dissolution of secondary phases, and radionuclide transport through engineered barriers and natural geologic barriers to the biosphere. Dakota is used to generate sets of representative realizations and to analyze parameter sensitivity.

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DOE OSTI2017

Vitrification of Sellafield Wastes with High Metallic Magnesium Alloy Content - 17495

The Sellafield nuclear complex, located in northwest England, is the largest such site in the United Kingdom. Sellafield Limited, which is the operating contractor for the Nuclear Decommissioning Authority, is evaluating thermal treatment technologies for a potential future facility to process a variety of Intermediate Level Waste sludges and slurries accumulated from historical site operations. In support of that effort, Atkins and the Vitreous State Laboratory (VSL) of The Catholic University of America conducted studies to evaluate the feasibility and implications of installing a joule-heated ceramic melter (JHCM) vitrification system to process a variety of sludge and slurry radioactive wastes at the Sellafield site. These sludges have the potential to contain a significant proportion of reactive metals in localized regions. The metallic fraction derives from unreacted Magnox spent fuel cladding. Thus, an important objective of this work was to investigate the performance of JHCM technology for the treatment of streams with localized regions of high reactive metal contents. The present work was focused on Site Ion Exchange Effluent Plant (SIXEP) Magnox sludges, investigated previously. Suitable waste simulants were developed for the SIXEP Magnox sludge waste stream, which included surrogates for all important radionuclides. High waste-loading, fully compliant glass formulations were designed using a combination of previous experience, extensive waste glass data bases, and small-scale crucible melts. Glasses from the test melts were characterized for properties that are important for processability and product quality. The most viable compositions were then selected for small-scale melter trials to collect information on processing rates and behavior, off-gas and secondary waste compositions, and overall system mass balance. These data were used to support the technology evaluation, which considered such factors as minimization of inactive additions, waste volume reduction, retention of radionuclides, minimization of secondary wastes, and overall process efficiency and effectiveness. The data were also used to develop sizing estimates for the full-scale facility and to estimate treatment durations under various treatment scenarios. A series of tests was then performed with partial replacement of the magnesium hydroxide with a metallic magnesium alloy that is close in composition to Magnox alloy. A commercially available magnesium alloy, Elecktron AZ31B, which has a nominal magnesium content of 95.7%, was used for this purpose. The testing was performed on a DuraMelter 10 (DM10) JHCM system to characterize the melter off-gas, glass product, and the effects of magnesium alloy particle size on feed processing. Data on processing rate and any processing issues, product glass composition, off-gas composition including hydrogen, and system operating parameters, including flows, temperatures, pressures, etc., were collected and analyzed to evaluate the effects of magnesium alloy on the vitrification process. (authors)

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DOE OSTI2021

Waste Compliance and Tracking System (WCATS) Version 3 Requirements Document

This document describes the end-user requirements for the Waste Compliance and Tracking System (WCATS) project in accordance with the WCATS Software Quality Management Plan, EPC-WMP-WCATSPLAN-001. The WCATS application shall support the generation, characterization, processing, and shipment of LANL radioactive, hazardous, and industrial waste. Regulatory drivers include RCRA hazardous waste, DOT shipping, NNSA nuclear material control and accountability, DOE nuclear safety, TSDF permit, and transuranic waste certification requirements. The system will utilize a task-based architecture that supports the spectrum of treatment, storage, disposal, administrative, and characterization based unit operations necessary to manage waste from cradle to grave. The application design shall readily accommodate new facilities, processes, workflow, signature requirements, and so forth, via end-user established metadata. WCATS will provide support for representing waste storage and disposal facilities, buildings, rooms, and grid layouts (x, y, z) to support waste and radioactive material inventory management. Nuclear material at risk (MAR), DOE hazard rating (e.g., Category II facility) compliance per DOE-STD-1027, and permit inventory requirements will be configurable for any storage or disposal facility, or waste operation, and the system will automatically evaluate and enforce those requirements. In addition, the application will support the characterization and management of the entire range of hazardous and radioactive wastes (TRU, MTRU, LLW, MLLW, hazardous waste, etc.) that might be colocated or processed at a permitted facility. Some capabilities not found in traditional systems include user-defined tank systems for liquid waste, user-defined work paths (i.e., sequence of operations), and an equipment subsystem for tracking the calibration, maintenance, and inspection of tools used to process waste, such as torque wrenches, scales, pH probes, etc. The application incorporates a desktop and mobile user interface as shown in Figure 1. The mobile interface supports field operations, such as waste item characterization, intra-facility transfers, internal and external audits, and shipment preparation and receipt.

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DOE OSTI2013

Treatment of spent solutions containing tributylphosphate

The realization of the process of treatment of organic radioactive waste (RW) containing tributylphosphate (TBP) in hydrocarbon solvent under the laboratory conditions is considered in this work. The main parameters of the process have been investigated as well as the characteristics of the final process product - sodium-aluminium-phosphate glass matrix. (authors)

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DOE OSTI2019

Pretreatment of Crystalline Silicotitanate (CST) and development of a digestion standard to support Tank Closure Cesium Removal (TCCR)

In support of Savannah River Remediation’s (SRR) deployment of an ion exchange process to remove radioactive cesium from waste supernate, referred to as the Tank Closure Cesium Removal (TCCR) process, SRNL was tasked with validating the field protocol for pretreating the crystalline silicotitanate (CST) material to be used in the columns, preparing CST for use in in-tank batch contact testing, as well as developing a digestion standard to verify complete dissolution of the CST prior to characterization after each in-tank batch contact test. A procedure (SRNL L29 Manual, ITS-0229) was developed to document the protocol for pretreating CST utilizing conditions similar to what will be performed in the field. Analysis of material pretreated following the procedure confirmed similar conversion to the Na+ form of the resin as had been obtained previously. Portions of the batch of material pretreated according to the developed protocol were also used for development of a digestion standard. Results from digestion and analysis of this standard were used to establish a set of control limits for the measured Ti, Zr, and Nb concentrations in digested CST material. An aliquot of this standard CST material will be digested and analyzed alongside each in-tank batch contact sample, and the measured concentrations of Ti, Zr, and Nb will be compared against the established control limits to confirm complete digestion has occurred.

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DOE OSTI2026

Assessment of Molybdate Based Corrosion Inhibitor for use at the Savannah River Tank Farm Cooling Water System

The Savannah River Site (SRS) has relied on 51 underground storage tanks, many dating back to the 1950s, to store radioactive liquid waste generated from nuclear processing and radionuclide production. These tanks use carbon steel cooling coils, cooled by soft, acidic water, to dissipate the decayed heat produced by radioactive waste. For decades, following historical industry practices, chromate has been added to the cooling water (~450 ppm at pH 9-11) to effectively inhibit corrosion, ensuring cooling coil longevity. Inspections of failed coils reveal that chromate-treated surfaces are generally well-preserved.

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DOE OSTI2013

Ab-initio simulations of chemical stability indicators of the bis-DGA-type molecule and its radiation degradation products

For hydrometallurgical treatment of the high level liquid waste (HLLW) in the DIAMEX and SANEX processes, organic compounds of the bis-DGA family are used as cation extractants in apolar solvents. For the compound of m-xylylene-bis-diglycolamide high distribution coefficients for Eu and Am were found. Since the environment of the process is highly radioactive and acidic (nitric acid), it is necessary to ensure the stability of the extractants. In order to analyse the process theoretically, the molecule of m-xylylene-bis- diglycolamide and two of its degradation products were simulated by the DFT computational methods (PBE, RPBE, BLYP, B3LYP) available within the simulation environment DMol{sup 3} 6.1 and Gaussian 09 software. The local chemical stability of some locations of the molecule was assessed from the calculated stability indicators (electrostatic potential, Fukui function, HOMO localization). In connection with the chemical treatment, especially the stability against an electrophilic attack was tested. The results of calculated bond orders and spatial distribution of electrostatic potential and HOMO were are successfully correlated with the local and general stability determined by the experiment. These results should be helpful for the further development of the separation process. (authors)

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DOE OSTI2018

Crystalline Silicotitanate Ion Exchange Column Sizing and Sensitivity Study in Support of the Hanford Test Bed Initiative

The Environmental Management Office of the Department of Energy (DOE) is spearheading a Test Bed Initiative a (TBI) to treat and dispose 2,000 gallons of tank waste from Hanford. One step in the process is to remove radioactive 137 Cs from the waste using Crystalline Silicotitanate b (CST) ion exchange media. This report describes the modeling performed to provide an estimate of the volume of a bed of CST required to meet the design requirements. Sensitivity studies were also performed to help provide project guidance. This modeling effort assumed 2100 gallons of waste to be processed (i.e., entire 2000 gallon campaign plus a 100 gallon margin).

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DOE OSTI2019

Cameco Corporation's Approach to Legacy and Ongoing Waste Management Within the Fuel Services Division - 19124

Since 2011, a Fuel Services Division (FSD) employee has been dedicated to oversee divisional waste management with the goal of accelerating the recycle/disposal of accumulated legacy radioactive wastes stored at its Ontario operations in Blind River and Port Hope. The Blind River Refinery (BRR) and Port Hope Conversion Facility (PHCF) were previously owned by the federal Crown Corporation, Eldorado Nuclear, becoming part of Cameco in 1988 and Cameco Fuel Manufacturing (CFM), previously known as Zircatec Precision Industries, becoming part of Cameco in 2006. Accumulated legacy radioactive waste is stored at each of the facilities. This radioactive waste has accumulated from operations beginning in the 1930's in Port Hope and the 1980's in Blind River. Due to minimal commercially available outlets for the disposal and/or treatment of these wastes, the waste was managed onsite by storage in underutilized space in buildings, trailers and/or in the secured yard. This relatively low-cost practice meant that over decades, the inventory increased at an unsustainable rate until available space became a limiting factor. During this time, the labelling, inventory records, characterization details and other history of many of the waste materials was not maintained adequately. Due to the volume of waste that had accumulated, campaigns to address these deficiencies in information were not considered feasible, sustaining the issue of accumulation of low-level radioactive waste. Since 2011, FSD has completed annual campaigns to address the accumulated legacy waste, which has included processing of material for decontamination to meet the unrestricted release criteria under the Canadian Nuclear Substance and Radiation Devices Regulations (NSRDR), disposal of marginally contaminated wastes (<500 ppm uranium) at a permitted landfill in the United States, processing of contaminated materials to reduce the volume of low-level radioactive waste and disposal at a designated facility in Port Hope. In addition FSD has undertaken significant changes to radioactive waste management practices at each of the facilities. Multiple characterization techniques, including visual observation, review of process and other historical records, laboratory analysis and radiation scanning have been utilized in campaign planning to identify the scope and appropriate waste management technique(s) to reduce the inventory of stored material. This divisional waste management program has resulted in a significant change to how radioactive waste is generated, inventoried, characterized and managed within FSD and has increased the visibility within the leadership team of the opportunities and challenges of waste management, such as alignment with decommissioning planning, project planning and budgeting. The heightened awareness has led to better financial and physical resource planning to execute waste projects with a current projected elimination of the legacy waste within FSD by 2025. (authors)

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DOE OSTI2017

An Efficient Tool-Set for Modeling Containment Ventilation in Very Complex Facilities - 17215

The applicability of a multi-zone ventilation analysis program, CONTAM, to complex radioactive waste material handling or processing plants is evaluated. Validation results are based on simulations from three-dimensional multi-physics computational fluid dynamics. Insight into the application limits of CONTAM and conditions under which the use of higher fidelity tools are required are also discussed. (authors)

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